Thermal-Hydraulic Analysis on the Encapsulated Nuclear Heat Source (ENHS)
Conference
·
OSTI ID:21160702
- Japan Nuclear Cycle Development Institute (JNC), 4002 Narita, O-arai, Ibaraki, 311-1393 (Japan)
- NESI, 4002 Narita, O-arai, Ibaraki, 311-1393 (Japan)
Thermal-hydraulic analysis was performed on the Encapsulated Nuclear Heat Source (ENHS). ENHS is a Lead-Bismuth cooled natural circulation fast reactor which is designed as a candidate Generation-IV reactor. The reactor has unique primary and secondary cooling systems flowed by natural circulation. In this study, the two-dimensional thermalhydraulic analysis method was applied to evaluate the basic cooling performance and flow distribution in the core. Core power profile in the radial direction was considered in the calculation to discuss the inherent flow distribution caused by buoyancy force. It was clarified that beyond 10% additional flow rate is automatically distributed to the hottest channel by the inherent flow distribution. In case of a ductless-core, the additional inherent flow distribution is reduced to 5%, because of the transverse flow inside the core. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21160702
- Country of Publication:
- United States
- Language:
- English
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